ML030160261

Deletion of Operational & Administrative Requirements from Yankee Nuclear Power Station Defueled TS

ML030160261
Person / Time
Site:	Yankee Rowe (DPR-003)
Issue date:	01/14/2003
From:	Heider K Yankee Atomic Electric Co
To:	Document Control Desk, NRC/FSME
References
-RFPFR, 268, BYR 2003-001
Download: ML030160261 (86)

Text

YANKEE ATOMIC ELECTRIC COMPANY 49 Yankee Road, Rowe, Massachusetts 01367 C7Y'AN4KEEk January 14, 2003 BYR 2003-001 P.C. No. 268 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

References:

(a)

License No. DPR-3 (Docket No. 50-29)

(b)

Letter, A. C. Kadak (YAEC) to T. E. Murley (USNRC), Permanent Cessation of Power Operations at the Yankee Nuclear Power Station, BYR 92-024, February 27, 1992.

(c)

Letter, M. B. Fairtile (USNRC) to J. M. Grant (YAEC), Issuance of Amendment #142 to Facility License DPR Yankee Nuclear Power Station, August 5, 1992.

(d)

Yankee Decommissioning Quality Assurance Program, Revision 30, August, 2000.

(e)

NRC Administrative Letter 95-06: "Relocation of Technical Specification Administrative Controls Related to Quality Assurance," dated December 12, 1995 (f)

NRC letter to Sacramento Municipal Utility District, Issuance of Amendment 129 to Facility Operating License No. DPR-54, dated February 5, 2002

Subject:

Deletion of Operational and Administrative Requirements from the Yankee Nuclear Power Station Defueled Technical Specifications Pursuant to 10 CFR 50.90 of the Commission's Rules and Regulations, Yankee Atomic Electric Company (YAEC) requests Nuclear Regulatory Commission (NRC) review and approval of a modification to the License and Appendix A of the Yankee Nuclear Power Station (YNPS)

Possession Only License (POL).

United States Nuclear Regulatory Commission P.C. No. 268 ATTN: Document Control Desk Page 2 PROPOSED CHANGES YAEC, in anticipation of transferring all of the YNPS spent nuclear fuel to dry cask storage within an Independent Spent Fuel Storage Installation (ISFSI), proposes the following License and Technical Specification changes:

1. Revise License Condition 2.C.(4) of the Possession Only License to remove reference to the Spent Fuel Pit. Revise License Condition 2.D to reflect continuation of the license in accordance with 10 CFR 50.51.

2. Delete the definitions from Technical Specification Section 1.0.

3. Delete the Limiting Conditions for Operation (LCOs), Surveillance Requirements and associated Bases from Technical Specification Section 3/4. LCOs 3/4.4 and 3/4.5 will be relocated to the Yankee Decommissioning Quality Assurance Program (YDQAP) with minor editorial changes to maintain consistency with other relocated text.

4. Revise Technical Specification Sections 5.1 "Site Location" to refer to the updated Final Safety Analysis Report, and 5.2 "Fuel Storage" to reflect how the YNPS spent fuel is being stored.

5. Revise Technical Specification Section 6.0 "Administrative Controls" to delete Sections 6.1 "Responsibility," 6.2 "Organization," and 6.4 "Training." Sections 6.3 "Facility Staff Qualifications, 6.10 "Radiation Protection Program," and 6.11 "High Radiation Area," will be relocated verbatim to the YDQAP. Sections 6.6 "Reportable Event Action, 6.8 "Reporting Requirements," 6.12 "Process Control Program," and 6.13 "Off-Site Dose Calculation Manual (ODCM)," will be relocated to the YDQAP with minor editorial changes to maintain consistency with other relocated text.

Table 1 "Regulatory Analysis Summary" identifies each License Condition or Technical Specification section to be deleted, relocated, or modified and provides the basis for the proposed change. An annotated version of the present TS pages showing the proposed changes is provided in Attachment I. The proposed new pages of the YNPS TS are given in Attachment II.

Attachment III provides a copy of the two LCOs and surveillance requirements as well as the seven administrative sections that are being relocated to the YDQAP.

United States Nuclear Regulatory Commission P.C. No. 268 ATTN: Document Control Desk Page 3 REASON AND BASIS FOR CHANGE YNPS shut down on October 1, 1991. During the outage and before February 26, 1992 all fuel assemblies, control rods, and neutron sources were removed from the Reactor Vessel and stored in the Spent Fuel Pit. On February 26, 1992, the YAEC Board of Directors decided to cease power operations permanently at YNPS. By letter, dated February 27, 1992 (Reference b),

YAEC notified the Nuclear Regulatory Commission (NRC) of the Company's decision to permanently cease power operations at the YNPS. On August 5, 1992, the NRC amended the YNPS Facility Operating License (DPR-3) to possession only status (Reference c).

In support of the decommissioning process currently underway at the YNPS, spent fuel is being transferred from the Spent Fuel Pit (SFP) to an Independent Spent Fuel Storage Installation (ISFSI) using casks certified for use under a general Part 72 license. Completion of YNPS decommissioning requires the eventual dismantlement of the Spent Fuel Pit. The proposed changes to delete operational and administrative requirements from the Technical Specifications will allow YAEC to eliminate those programs and activities no longer necessary to ensure safe storage of spent fuel within the SFP following transfer of all spent fuel to the ISFSI.

Furthermore, continuing to implement these requirements would cause YAEC to incur significant expense with no commensurate benefit to protecting public health and safety. The continued safe storage of the spent fuel is ensured by YAEC's adherence to the license conditions and technical specifications contained within the dry cask system's Certificate of Compliance.

YAEC proposes to revise Possession Only License Condition 2.C.(4). YAEC continues to comply with the fire protection requirements of 10 CFR 50.48(f) for those licensees who have already submitted their §50.82(a)(1) certifications as well as those fire protection requirements necessary to satisfy the dry cask licensing basis as documented in the 72.212 evaluation.

However, part of this License Condition addresses the ability to maintain the fuel in the Spent Fuel Pit in a safe condition in the event of a fire. Following the complete transfer of all spent nuclear fuel from the Spent Fuel Pit to the Independent Spent Fuel Storage Installation (ISFSI),

this portion of the license condition will no longer apply. Because fire protection requirements for permanently shutdown plants are already contained in regulation and since part of the license condition concerning the Spent Fuel Pit will no longer be applicable, revision of License Condition 2.C.(4) to remove reference to the Spent Fuel Pit is acceptable.

YAEC also proposes to revise License Condition 2.D. Currently, this License Condition states that the amended license would expire at midnight, July 9, 2000. The proposed text reflects that the license continues until terminated by the Nuclear Regulatory Commission as stated in 10 CFR 50.51 and is therefore acceptable.

United States Nuclear Regulatory Commission P.C. No. 268 ATTN: Document Control Desk Page 4 The NRC established the regulatory requirements for the content of Technical Specifications in 10 CFR 50.36 "Technical Specifications." In 10 CFR 50.36, the emphasis is placed on matters related to the prevention of accidents, mitigation of accident consequences, and items directly related to maintaining the integrity of the physical barriers designed to contain radioactivity.

Specifically, 10 CFR 50.36 requires that Technical Specifications include the following five specific categories:

1. Safety limits, limiting system settings, and limiting control settings. [10 CFR 50.36(c)(1)]

2. Limiting conditions for operation. [10 CFR 50.36(c)(2)]

3. Surveillance requirements. [10 CFR 50.36(c)(3)]

4. Design features. [10 CFR 50.36(c)(4)]

5. Administrative controls. [10 CFR 50.36(c)(5)]

The regulation does not specify explicit items to be included in Technical Specifications. As detailed in the Regulatory Analysis Summary (Table 1), YAEC proposes to delete the definitions, delete the LCOs, Surveillance Requirements and associated Bases, revise the text for Design Feature "Fuel Storage," and delete/relocate certain Administrative Controls.

10 CFR 50.36(c)(1), Safety limits, limiting system settings, and limiting control settings This section is no longer applicable to the Yankee Nuclear Power Station and is not part of the current YNPS Defueled Technical Specifications.

10 CFR 50.36(c)(2), Limiting conditions for operation As stated in 10 CFR 50.36(c)(2)(ii), a technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

Criterion 1 - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3 - A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4 - A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

United States Nuclear Regulatory Commission P.C. No. 268 ATTN: Document Control Desk Page 5 Of the five LCOs currently in the YNPS Defueled Technical Specifications, three are directly related to the storage of spent nuclear fuel in the spent fuel pit. After all the spent nuclear fuel has been transferred to the ISFSI, none of the four criteria listed above would apply and the LCOs would no longer be applicable. Therefore, it is acceptable to delete these LCOs from the Technical Specifications.

The other two LCOs implement requirements already contained in regulations 10 CFR 20 Appendix B and 10 CFR 70.39(c). As discussed in Table 1, YAEC will continue to implement the surveillance procedures used to ensure compliance with these regulations. Since the surveillance program is implemented through other licensee-controlled documents, it is acceptable to delete these two LCOs from the Technical Specifications.

10 CFR 50.36(c)(3), Surveillance requirements Surveillance requirements relate to testing, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that the limiting conditions for operation will be met.

As discussed above, following transfer of all the spent nuclear fuel to the ISFSI, the LCOs are no longer applicable or controlled in other documents. Deleting the LCOs eliminates the need for surveillance requirements in the Technical Specifications. Therefore, the proposed deletion of the surveillance requirements is acceptable.

10 CFR 50.36(c)(4), Design features Design features are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in subparagraphs (c)(1), (2), and (3) of this section.

Technical Specification 5.1 currently identifies the plant location and provides the geographical coordinates of the reactor containment structure. The proposed amendment revises this text to refer to a figure contained within the updated Final Safety Analysis Report. Reference to an FSAR will allow site conditions to be more accurately reflected as structures such as containment are removed as part of decommissioning. Therefore the proposed revision is acceptable.

Technical Specification 5.2 currently addresses spent fuel pit design characteristics, storage capacity, and criticality control. The proposed amendment deletes the text of this section and replaces it with the following:

"A maximum of 540 spent fiel assemblies from the Yankee Nuclear Power Station are stored in dry casks within an Independent Spent Fuel Storage Installation (ISFSI). "

United States Nuclear Regulatory Commission P.C. No. 268 ATTN: Document Control Desk Page 6 The text revision is intended to clearly indicate how the spent nuclear fuel is being stored. With no spent fuel being stored in the Spent Fuel Pit, there is no longer a concern regarding criticality, leakage or capacity since these design features are no longer applicable. Therefore, the proposed revision is acceptable.

10 CFR 50.36(c)(5), Administrative controls Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

As stated in NRC's Safety Evaluation related to Amendment No. 129 to Rancho Seco Facility Operating License No. DPR-54 (Reference f), the particular controls to be included in the Technical Specifications are the provisions that the Commission deems essential for the safe operation of the facility that are not already covered by other regulations. The staff determined that administrative control requirements that are not specifically required under 10 CFR 50.36(c)(5), and which are not otherwise necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety, may be relocated to more appropriate documents (e.g., quality assurance program), which are subject to regulatory change control.

The quality assurance program is a logical candidate for the relocation of administrative controls because the controls imposed by 10 CFR 50 Appendix B, the existing NRC-approved quality assurance plans and commitments to industry quality assurance standards, and the established quality assurance program change control process in 10 CFR 50.54(a) are maintained.

NRC Administrative Letter (AL) 95-06 "Relocation of Technical Specification Administrative Controls Related to Quality Assurance" provides guidance to licensees requesting amendments that relocate administrative controls to NRC-approved quality assurance programs where changes are controlled in accordance with 10 CFR 50.54(a).

As discussed in Table 1, deleting the administrative controls proposed in the amendment is consistent with the guidance in AL 95-06 and is therefore acceptable.

Furthermore, as stated in the July 22, 1993 "Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" (58 FR 39132), the Commission noted that in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be retained in the Technical Specifications, it was adopting the qualitative standard enunciated by the Atomic Safety and Licensing Board in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979). In this proceeding, the Appeal Board observed:

United States Nuclear Regulatory Commission P.C. No. 268 ATTN: Document Control Desk Page 7 There is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally binding upon the licensee unless and until changed with specific Commission approval. Rather, as best we can discern it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

Lastly, 10 CFR 50.36(c)(6) "Decommissioning" states that for nuclear power reactor facilities that have submitted the certifications required by §50.82(a)(1), Technical specifications involving limiting conditions for operation; surveillance requirements; design features; and administrative controls will be developed on a case-by-case basis.

Based on the discussion above, YAEC believes that the proposed amendment is completely consistent with the content of NRC Administrative Letter 95-06 as well as the Commission Policy cited above and that the health and safety of the public will not be endangered by operation of the facility in the proposed manner.

SIGNIFICANT HAZARDS CONSIDERATION YAEC has reviewed the proposed amendment against each of the criteria in 10 CUR 50.92 and has concluded that the amendment request involves no significant hazards consideration. The following discussion provides YAEC's response to the no significant hazards consideration criteria:

1. Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The proposed changes reflect the complete transfer of all spent nuclear fuel from the Spent Fuel Pit to the Independent Spent Fuel Storage Installation (ISFSI). Design basis accidents related to the Spent Fuel Pit are discussed in the YNPS FSAR. These postulated accidents are predicated on spent nuclear fuel being stored in the Spent Fuel Pit. With the removal of the spent fuel from the Spent Fuel Pit, there are no remaining important to safety systems required to be monitored and there are no remaining credible accidents that require that actions of a Certified Fuel Handler or non-Certified Fuel Handler to prevent occurrence or mitigate the consequences.

The YNPS FSAR provides a discussion of radiological events postulated to occur as a result of decommissioning with the bounding consequence resulting from a materials handling event. The proposed changes do not have an adverse impact on decommissioning activities or any of their postulated consequences.

United States Nuclear Regulatory Commission P.C. No. 268 ATTN: Document Control Desk Page 8 The proposed change to the Design Features section of the Technical Specifications clarifies that the spent fuel is being stored in dry casks within an ISFSI. The probability or consequences of accidents at the ISFSI are evaluated in the dry cask vendor's FSAR and are independent of the accidents evaluated in the YNPS FSAR.

Based on the above, the proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed license amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

No. The proposed changes reflect the reduced operational risks as a result of the spent nuclear fuel being transferred to dry casks within an ISFSI. The proposed changes do not modify any physical systems, or components. The plant conditions for which the YNPS FSAR design basis accidents relating to spent fuel have been evaluated are no longer applicable. The aforementioned proposed changes do not affect any of the parameters or conditions that could contribute to the initiation of an accident. Design basis accidents associated with the dry cask storage of spent fuel are already considered in the dry cask system's Final Safety Analysis Report. No new accident scenarios are created as a result of deleting non-applicable operational and administrative requirements. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed license amendment involve a significant reduction in a margin of safety?

No. As described above, the proposed changes reflect the reduced operational risks as a result of the spent nuclear fuel being transferred to dry casks within an ISFSI. The design basis and accident assumptions within the YNPS FSAR and the Defueled Technical Specifications relating to spent fuel are no longer applicable. The proposed changes do not affect remaining plant operations, systems, or components supporting decommissioning activities. In addition, the proposed changes do not result in a change in initial conditions, system response time, or in any other parameter affecting the course of a decommissioning activity accident analysis. Therefore, the proposed changes will not involve a significant reduction in the margin of safety.

Based on the considerations noted above, it is concluded that the proposed changes will not endanger the public health and safety.

United States Nuclear Regulatory Commission P.C. No. 268 ATTN: Document Control Desk Page 9 ENVIRONMENTAL IMPACT DETERMINATION This amendment request meets the criteria specified in 10 CFR 51.22 (c)(9) for categorical exclusion or otherwise not requiring environmental review. Specific criteria contained in this section of the regulations are discussed below:

1. The amendment involves no significant hazards consideration. As demonstrated above, this requested amendment does not involve any significant hazards considerations.

2. There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite. The amendment removes operational and administrative requirements for systems that are no longer functionally required to support the safe storage of spent nuclear fuel within the Spent Fuel Pit. The proposed changes are administrative in nature and do not affect any systems such that there may be an increase or change in type of effluents discharged offsite.

3. The deletion of the aforementioned license condition as well as the elimination of non-applicable operational and administrative requirements from the Technical Specifications will not result in a significant increase in individual or cumulative occupational radiation exposure.

Based on the foregoing, it is concluded that the proposed amendment meets the criteria for categorical exclusion set forth in 10 CFR 51.22 (c)(9) and therefore, no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

ISR AND IRAC REVIEW The proposed changes have received an Independent Safety Review (ISR) as well as a review by the Independent Review and Audit Committee (IRAC) and have been determined to be appropriate.

United States Nuclear Regulatory Commission P.C. No. 268 ATTN: Document Control Desk Page 10 SCHEDULE OF CHANGE The proposed changes, following approval by the Commission, will be implemented within 60 days from the transfer of the last spent fuel cask out of the Spent Fuel Pit. The current dry cask load schedule projects that YAEC will need NRC completion of this licensing action by April 1, 2003.

Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY Kenneth J. Heider Vice President Attachments Table I - Regulatory Analysis Summary Attachment I - Annotated Version of License and Technical Specification Pages Attachment II - Clean Version of License and Technical Specification Pages 1I1-Text Relocated to the YDQAP C:

J. Hickman, USNRC, Project Manager R. Bellamy, USNRC, Region I STATE OF CONNECTICUT MIDDLESEX COUNTY Then personally appeared before me, Kenneth J. Heider, who, being duly sworn, did state that he is Vice President of Yankee Atomic Electric Company, that he is duly authorized to execute and file the foregoing document in the name and on behalf of Yankee Atomic Electric Company, and that the statements therein are true to the best of his knowledge and belief.

Gerard van Noordennen Notary Public My Commission Expires December 31, 2003

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section License Condition 2.C.(4)

The latter portion of this fire protection license condition addresses the ability to maintain the fuel in the Spent Fuel Pit in a safe condition in the event of a fire. Following the complete transfer of all the spent nuclear fuel to the Independent Spent Fuel Storage Installation (ISFSI), this portion of the license condition is no longer applicable. Therefore, revision of this license condition to remove the reference to the Spent Fuel Pit is acceptable.

License Condition 2.D This license condition states that the amended license expired at midnight, July 9, 2000. However, pursuant to 10CFR50.51, the license actually continues until it is terminated by the NRC. The revised license condition text more accurately reflects the regulatory requirements for license termination and is therefore acceptable.

Definitions As discussed below, with the complete transfer of spent fuel from the Spent Fuel Pit (SFP) to the ISFSI, there will no longer be any Limiting Conditions for Operations (LCOs). As such, the definitions of ACTION, CHANNEL CALIBRATION, CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and OPERABLE - OPERABILITY are no longer applicable.

The terms MEMBER(S) OF THE PUBLIC and REPORTABLE EVENT are defined in other regulations, including 10 CFR Part 20 and 10 CFR 50.73 respectively.

The Administrative Sections addressing the Off-Site Dose Calculation Manual (ODCM) as well as the Process Control Program (PCP) are being relocated to the YDQAP. The definitions of ODCM and PCP are adequately described within each of the manuals. Consequently, the terms OFF-SITE DOSE CALCULATION MANUAL (ODCM) and PROCESS CONTROL PROGRAM (PCP) are no longer applicable.

These definitions are either no longer applicable, or are redundant with definitions in other regulations. Therefore, deleting these definitions from the Technical Specifications are acceptable.

Page 1 of 15

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section Limiting Conditions for Each subsection as discussed below, following the complete Operation and Surveillance transfer of spent fuel from the SFP to the ISFSI, may be deleted Requirements from the Technical Specifications or relocated to the Yankee Defueled Quality Assurance Program (YDQAP).

3/4.0 Applicability Specifications 3.0.1, 3.0.2, 3.0.3, and 3.0.4 establish the general 3.0.1, 3.0.2, 3.0.3, and 3.0.4 requirements applicable to LCOs.

Limiting Conditions for Operation As discussed below, after all of the spent fuel is transferred from the SFP to the ISFSI, the Section 3/4 LCOs are no longer required, thereby eliminating the need for LCOs 3.0.1, 3.0.2, 3.0.3, and 3.0.4. Consequently, deleting LCOs 3.0.1, 3.0.2, 3.0.3, and 3.0.4 are acceptable.

4.0.1, 4.0.2, 4.0.3, and 4.0.4 Surveillance Requirements ensure that conditions specified by the Surveillance Requirements LCO are met. Technical Specification Sections 4.0.1, 4.0.2, 4.0.3, and 4.0.4 specify the conditions when surveillances are required, the required frequency for surveillances, and what constitutes non compliance with LCO operability requirements.

As discussed below, the Section 3/4 surveillance requirements are to be deleted after all of the spent fuel is transferred from the SFP to the ISFSI thereby eliminating the need for surveillance requirements 4.0.1, 4.0.2, 4.0.3, and 4.0.4. Consequently, deleting surveillance requirements 4.0.1, 4.0.2, 4.0.3, and 4.0.4 are acceptable.

Page 2 of 15

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section 3/4.1 Technical Specification 3/4.1 provides the LCO, Applicability, Spent Fuel Pit Water Level Action and Surveillance Requirements for spent fuel pit water level. This section requires that at least 14 feet of water be maintained over the top of the irradiated fuel assemblies when those assemblies are seated in their racks and at least 5 feet of water be maintained over the top of the irradiated fuel assemblies during fuel handling operations unless the assemblies are in a shipping or transfer cask. This section also specifies the frequency that the SFP water level is to be verified. The Applicability for the LCO states:

"Whenever irradiated fuel assemblies are in the spent fuel pit."

After all of the spent fuel is transferred from the SFP to the ISFSI, the applicability condition will have been permanently removed and the need to maintain or monitor a specified water level in the spent fuel pit will no longer exist. Therefore, the proposed deletion is acceptable.

3/4.2 Technical Specification 3/4.2 provides the LCO, Applicability, Crane Travel - Spent Fuel Pit Action and Surveillance Requirements for heavy loads traveling over the spent fuel pit. This section requires that loads in excess of 900 lbs be prohibited from traveling over the spent fuel pit unless the load fits into one of ten exempted equipment categories.

The Applicability for the LCO states:

"With fuel assemblies in the Spent Fuel Pit."

After all of the spent fuel is transferred from the SFP to the ISFSI, the applicability condition will have been permanently removed and the need to maintain the crane travel restriction over the spent fuel pit will no longer exist. Therefore, the proposed deletion is acceptable.

Page 3 of 15

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section 3/4.3 Technical Specification 3/4.3 provides the LCO, Applicability, Spent Fuel Storage Area Action and Surveillance Requirements for monitoring radiation Radiation Monitor levels in the spent fuel storage area. This section requires that radiation levels in the spent fuel storage area be monitored by a fixed radiation monitor. This section also specifies the frequency that the radiation monitor be tested. The Applicability for the LCO states:

"When handling irradiated fuel, control rods or sources."

After all of the spent fuel, control rods or sources are transferred from the SFP to the ISFSI, the applicability condition will have been permanently removed and the need to monitor radiation levels in the spent fuel storage area will no longer exist.

Therefore, the proposed deletion is acceptable.

3/4.4 Technical Specification 3/4.4 provides the LCO, Applicability, Liquid Hold-Up Tanks Action and Surveillance Requirements for restricting the quantity of radioactive material contained in outside tanks. This section requires that the quantity of radioactive material contained in tanks be limited to less than or equal to 10 Curies, excluding tritium and dissolved or entrained noble gases. This section also specifies the frequency that the material contained in each tank be analyzed.

Restricting the quantity of radioactive material contained in outside tanks provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting concentration at the nearest potable water supply and the nearest surface water supply in an unrestricted area would be less than the limits of 10 CFR 20, Appendix B.

Although the proposed amendment deletes this LCO from the Technical Specifications, the requirement to verify that the quantity of radioactive material contained in tanks be limited to less than or equal to 10 Curies, excluding tritium and dissolved or entrained noble gases, will be relocated and maintained in the Yankee Decommissioning Quality Assurance Program (YDQAP) with minor editorial changes to remove reference to another Page 4 of 15

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section section also being relocated.

3/4.4 (Continued)

NRC Administrative Letter 95-06 provides a discussion concerning the relocation of Technical Specification administrative controls to a quality assurance program. The NRC considers relocating these requirements to the quality assurance program as logical because of the controls imposed by 10 CFR 50 Appendix B, the existence of an NRC approved quality assurance program, and the quality assurance program change control process in 10 CFR 50.54(a).

Furthermore, as stated in the July 22, 1993 "Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" (58 FR 39132), the Commission noted that in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be retained in the Technical Specifications, it was adopting the qualitative standard enunciated by the Atomic Safety and Licensing Board in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979). In this proceeding, the Appeal Board observed:

There is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally binding upon the licensee unless and until changed with specific Commission approval. Rather, as best we can discern it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

Since the intent of this LCO is to meet limits already specified in 10 CFR 20 Appendix B and the administrative requirements will be maintained in the YDQAP, the proposed relocation of the aforementioned requirements is completely consistent with the Commission Policy cited above as well as with the content of NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

Page 5 of 15

Table 1 - Reiulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section 3/4.5 Sealed Source Contamination Technical Specification 3/4.5 provides the LCO, Applicability, Action and Surveillance Requirements for meeting the requirements of 10 CFR 70.39 (c) "Specific licenses for the manufacture or initial transfer of calibration or reference sources" and ensures that leakage from special nuclear material sources will not exceed allowable intake values. This section requires sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material, or 5 microcuries of alpha emitting material, shall be free of > 0.005 microcuries of removable contamination. This section also specifies the frequency that each sealed source will be tested for leakage and/or contamination.

Although the proposed amendment deletes this LCO from the Technical Specifications, the requirement to test each sealed source for leakage and/or contamination will be relocated and maintained in the Yankee Decommissioning Quality Assurance Program (YDQAP) with minor editorial changes to remove reference to another section also being relocated.

Since the intent of this LCO is to meet limits already specified in 10 CFR 70.39(c) and the administrative requirements will be maintained in the YDQAP, the proposed relocation of the aforementioned requirements is completely consistent with the Commission Policy cited earlier as well as with the content of NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

3/4 Bases This proposed amendment deletes the Bases section in its entirety.

This is consistent with the deletions discussed above, and is therefore acceptable. The proposed amendment also revises the Index to reflect the proposed changes. This is consistent with the proposed changes and is therefore acceptable.

Page 6 of 15

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section 5.0 Design Features 5.1 Site Location Technical Specification 5.1 currently identifies the location of the Yankee Nuclear Power Station and provides the geographical coordinates of the reactor containment structure. The proposed amendment revises this text to refer to a figure contained within the updated Final Safety Analysis Report. Reference to an FSAR figure more accurately reflects site conditions as the existing referenced containment structure is scheduled to be removed as part of the decommissioning process. Therefore, the proposed revision is acceptable.

5.0 Design Features 5.2 Fuel Storage Technical Specification 5.2 currently addresses spent fuel pit design characteristics, storage capacity, and criticality control.

The proposed amendment deletes the text of this section and replaces it with the following:

"The spent fuel assemblies from the Yankee Nuclear Power Station are stored in 15 dry casks within an Independent Spent Fuel Storage Installation (ISFSI) with a maximum capacity of 540 spent fuel assemblies. "

The text revision is intended to clearly indicate how the spent nuclear fuel is being stored. With no spent fuel being stored in the Spent Fuel Pit, there is no longer a concern regarding criticality, leakage or capacity. Therefore, the proposed revision is acceptable.

Page 7 of 15

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section 6.0 Administrative Controls 6.1 Responsibility 6.1.1 6.1.2 The proposed amendment deletes Administrative Controls Sections 6.1, 6.2, and 6.4,. As discussed below, these administrative requirements are no longer applicable with all of the spent fuel stored at the ISFSI. The remaining administrative controls, 6.3, 6.6, 6.8, 6.10, 6.11, 6.12 and 6.13 can be relocated to the YDQAP.

NRC Administrative Letter 95-06 provides a discussion concerning the relocation of Technical Specification administrative controls to a quality assurance program. The NRC considers relocating these requirements to the quality assurance program as logical because of the controls imposed by 10 CFR 50 Appendix B, the existence of an NRC approved quality assurance program, and the quality assurance program change control process in 10 CFR 50.54(a).

Relocating Technical Specification administrative controls to the NRC-approved YDQAP may reduce resources spent by both Yankee and the NRC in processing license amendment requests.

Furthermore, many administrative controls pertained to the safe storage of irradiated fuel within the Spent Fuel Pit. With the transfer of spent nuclear fuel to the ISFSI, these administrative controls are no longer applicable.

Technical Specification 6.1.1 provides a general, one-sentence description of the Decommissioning Manager (Site Manager) stating that the Decommissioning Manager is responsible for the overall operation of the facility. Technical Specification 6.1.2 states that in this capacity, the Decommissioning Manager reports, and is responsible to, the Vice President of YAEC.

Although not identically worded, the YDQAP,Section I "Organization" provides an equivalent description of the Decommissioning Manager's responsibilities and reporting lines.

Providing the responsibility description in the YDQAP is consistent with the guidance in NRC Administrative Letter 95-06.

Page 8 of 15

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section Therefore, deleting the organizational responsibility controls from 6.1 Responsibility 6.1 is acceptable because an equivalent description already exists (Continued) in the YDQAP. Moreover, similar changes were approved as amendment 129 for Rancho Seco in February of 2002 (Reference f).

6.2 Organization Technical Specification 6.2.1 provides a general discussion of the 6.2.1 Onsite and Offite corporate and onsite organizations established for corporate Organizations management and safe facility operations so that nuclear safety is assured.

As discussed above, the YDQAP,Section I "Organization" provides an equivalent description of the corporate and onsite organizations responsible for the overall safe operation of the facility. In addition, the YNPS FSAR Section 502 "Decommissioning Organization and Responsibilities" provides a discussion of YAEC's commitment, goals and organizational strategy and is consistent with the organization described in the YDQAP as well as Technical Specification 6.2.1.

Providing the corporate and onsite organizational description in the YDQAP is consistent with the guidance in NRC Administrative Letter 95-06. Therefore, deleting the organizational description from 6.2.1 is acceptable because an equivalent description already exists in the YDQAP and the YNPS FSAR. Similar changes were approved as amendment 129 for Rancho Seco in February of 2002 (Reference f).

Page 9 of 15

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section Page 10 of 15 6.2.2 Facility Organization I

Subparagraph "a" established the minimum shift crew composition. The required complement was one Certified Fuel Handler and one Equipment Operator. With removal of all the spent fuel from the Spent Fuel Pit, there are no remaining important to safety systems required to be monitored. Therefore, the requirement to maintain a Certified Fuel Handler and Equipment Operator onsite, at least one of whom is located in the control room, is no longer necessary. Therefore, it is acceptable to delete this requirement and Table 6.2-1.

Subparagraph "b" is applicable only when fuel is in the Spent Fuel Pit. With transfer of all the spent fuel from the Spent Fuel Pit to the ISFSI, this is no longer necessary and it is acceptable to delete this requirement.

Subparagraph "c" is applicable only during fuel handling operations. With transfer of all the spent fuel from the Spent Fuel Pit to the ISFSI, there will no longer be any fuel handling operations. Therefore, it is acceptable to delete this requirement.

Subparagraph "d" is applicable only when fuel is in the Spent Fuel Pit. With transfer of all the spent fuel from the Spent Fuel Pit to the ISFSI, there will be no functions associated with the safe storage of fuel within the Spent Fuel Pit. Therefore, it is acceptable to delete this requirement.

Subparagraph "e" is associated with the supervision of fuel handling operations by a Certified Fuel Handler. With transfer of all the spent fuel from the Spent Fuel Pit to the ISFSI, there will no longer be any fuel handling operations. Therefore, it is acceptable to delete this requirement.

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section 6.3 Facility Staff This section specifies the minimum qualifications that the facility Qualifications management, supervisory staff, and the Radiation Protection Manager shall meet.

Although the proposed amendment deletes this administrative section from the Technical Specifications, the facility staff qualifications will be relocated verbatim and maintained in the Yankee Decommissioning Quality Assurance Program (YDQAP).

Since this administrative requirement will be maintained in the YDQAP, the proposed relocation of the aforementioned requirements is completely consistent with the Commission Policy cited earlier as well as with the content of NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

6.4 Training This section specifies the training requirements for the Certified Fuel Handlers as being conducted in accordance with an NRC approved program and the training requirements for the unit staff as meeting the requirements of ANSI Ni18.1-1971 Section 5.5.

However, the training requirements of ANSI N18.1-1971 Section 5.5 are oriented to plant operating events such as startup, operational limitations, and abnormal operating procedures. With transfer of all the spent fuel from the Spent Fuel Pit to the ISFSI, there will no longer be any fuel handling operations. Therefore, these requirements no longer apply. In addition, with there no longer being the need for Certified Fuel Handlers, their training program is no longer applicable. Therefore, it is acceptable to delete these requirements.

Page 11 of 15

Table 1 - Re2ulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section 6.6 Reportable Event This section specifies that each reportable event shall be submitted Action in accordance with 10 CFR 50.73 and that each event shall receive an Independent Safety Review and be submitted to the Independent Review and Audit Committee (IRAC) as well as the Decommissioning Manager.

Although the proposed amendment deletes this administrative section from the Technical Specifications, the reportable event requirements will be relocated and maintained in the YDQAP with a minor editorial change. The title "Decommissioning Manager" will be changed to "Site Manager." This is a title change only and has no impact on the roles and responsibilities of the individual holding this position.

Since the intent of this administrative section is to meet limits already specified in 10 CFR 50.73 and the administrative requirements will be maintained in the YDQAP, the proposed relocation of the aforementioned requirements is completely consistent with the Commission Policy cited earlier as well as with the content of NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

6.8 Reporting This section specifies that reports shall be submitted in accordance Requirements with 10 CFR 50.4 and Regulatory Guide 1.16 and provides the due dates and regulatory basis for each report.

6.8.1 6.8.2 Although the proposed amendment deletes this administrative 6.8.3 section from the Technical Specifications, the requirements for submittal of the specified reports will be relocated and maintained in the YDQAP with minor editorial changes. The references to the reporting requirements of 6.8.1, 6.8.2, and 6.8.3 within this section have been revised to maintain consistency.

In addition, 10CFR70.39(c) specifies detection limits and the manner in which sealed sources licensed under Part 70 are to be wipe tested to determine whether they are leaking or losing plutonium. Specification 3/4.5 is based on these limits to ensure that leakage from by-product, source and special nuclear material Page 12 of 15

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section sources will not exceed allowable intake values. In order to 6.8 (Continued) maintain consistency with the proposed deletion and relocation of Technical Specification 3/4.5 to the YDQAP, the reference in 6.8.3.a to Specification 3.5 is editorially revised to 10 CFR 70.39 (c) so that 6.8.3.a reads as follows:

"Sealed Source leakage in excess of 10 CFR 70.39 (c) limits."

Since the intent of this administrative section is to comply with reporting requirements already specified in 10 CFR 20.2206 and 10 CFR 50 Appendix I and since the administrative requirements will be maintained in the YDQAP, the proposed relocation of the aforementioned requirements is completely consistent with the Commission Policy cited earlier as well as with the content of NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

6.10 Radiation Protection This section specifies that procedures for personnel radiation Program protection shall be consistent with 10 CFR 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposures.

Although the proposed amendment deletes this administrative section from the Technical Specifications, the Radiation Protection Program requirements will be relocated verbatim and maintained in the YDQAP.

Since the intent of this administrative section is to meet limits already specified in 10 CFR 20 and the administrative requirements will be maintained in the YDQAP, the proposed relocation of the aforementioned requirements is completely consistent with the Commission Policy cited earlier as well as with the content of NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

Page 13 of 15

Table 1 - Regulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section 6.11 High Radiation Area This section specifies the requirements for the posting and control of access to High Radiation Areas.

Although the proposed amendment deletes this administrative section from the Technical Specifications, the High Radiation Area requirements will be relocated verbatim and maintained in the YDQAP.

Since the intent of this administrative section is to meet requirements already specified in 10 CFR 20 Subpart G and Subpart J and the administrative requirements will be maintained in the YDQAP, the proposed relocation of the aforementioned requirements is completely consistent with the Commission Policy cited earlier as well as with the content of NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

6.12 Process Control This section specifies how to prepare, review, approve and retain Program changes to the Process Control Program (PCP) for on-site processing and packaging of wet solid radioactive wastes.

Although the proposed amendment deletes this administrative section from the Technical Specifications, the Process Control Program requirements will be relocated and maintained in the YDQAP with a minor editorial change. The title "Decommissioning Manager" will be changed to "Site Manager."

This is a title change only and has no impact on the roles and responsibilities of the individual holding this position.

Since the intent of this administrative section is to comply with requirements already contained in 10 CFR, Parts 20, 61 and 71 and since the administrative requirements will be maintained in the YDQAP, the proposed relocation of the aforementioned program requirements is completely consistent with the Commission Policy cited earlier as well as with the content of NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

Page 14 of 15

Table 1 - Reeulatory Analysis Summary License / Appendix A Description of the Proposed License and Technical Specification Technical Specification Changes Section 6.13 Off-Site Dose This section specifies how to prepare, review, approve, and submit Calculation Manual changes to the ODCM.

(ODCM)

Although the proposed amendment deletes this administrative section from the Technical Specifications, the ODCM change requirements will be relocated and maintained in the YDQAP with a minor editorial change. The title "Decommissioning Manager" will be changed to "Site Manager." This is a title change only and has no impact on the roles and responsibilities of the individual holding this position.

Since the intent of this administrative section is to meet limits already specified in 40 CFR 190, 10 CFR 20, 10 CFR 50.36(a),

and 10 CFR 50 Appendix I and since the administrative requirements will be maintained in the YDQAP, the proposed relocation of the aforementioned section is completely consistent with the Commission Policy cited earlier as well as with the content of NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

Page 15 of 15

ATTACHMENT I Revised License and Technical Specification Pages Annotated Version License 3

5 Appendix A i through v 1-1 1-2 3/4-1 through 3/4-8 Bases Cover Page B3f4-1 through B3/4-7 5-1 6-1 through 6-5 6-13 6-16 through 6-22

Amendment "No. 145 914/1992 and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or Incorporated below:

(1)

Maximum Power Level The licensee Is not authorized to operate the reactor. Fuel may not be placed in the reactor vessel.

(2)

Technical Specifications The Technical Specifications contained In Appendix A, as revised through Amendment No. 155, are hereby Incorporated in the Ucense. The licensee shall possess and maintain the facility In accordance with the Technical Specifications.

(3)

Physical Protection Amendment No. 156 3/1132002 The licensee shall fully Implement and maintain In effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans Including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revision to 10 CFR 73.55 (51 FR 27817 and 17822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plan, which contain Safeguards Information protected under 10 CFR 73.21, is entitled, "Yankee Nuclear Power Station Security Plan,"

which Includes the "Contingency Plan' and the "Guard Training and Qualification Plan," with revisions submitted through June 28, 2001.

Changes made In accordance with 10 CFR 73.55 shall be Implemented In accordance with the schedule set forth therein.

(4)

Fire Protection Amendment No. 144 8/20/1992 The licensee shall Implement and maintain in effect all provisions of the approved Fire Protection Program as described In the Final Safety Analysis Report for the facility and as approved by NRC Safety Evaluation Reports dated March 15, 1979, and as supplemented October 1, 1980, and August 27, 1986, subject to the following provisions:

The licensee may make changes to the approved Fire Protection Program without prior NRC approval only if thoce changec would not adVOrcolY affOot the ability to maintain the fuel in the Spent Fuol Pit in a safe o*ndition in thoe vn.t _1 a, firoif these chanaies do not reduce the effectiveness of fire protection for facilities, wvstems. and eguioment which could result in a radiological hazard, taking Into account the decommissioning Wlant conditions and activities.

D.

T-his; amendod licnso is offectiv as of the date-of iccuanco and Eshall oxpiro at midnightv July 0,-20OQ.This license is effective as of the date of issuance and authorizes ownership and Possession of this facility until the Commission notifies the licensee in writing that the license is terminated. The licensee shall:

1. Take actions necessary to decommission and decontaminate this facility and continue to maintain this facility, including, where applicable, the storage, control and maintenance of the spent fuel, In a safe condition: and

2. Conduct activities in accordance with all other restrictions applicable to this facility in accordance with NRC regulations and the specific provisions of this 100FR50 facility license.

FOR THE NUCLEAR REGULATORY COMMISSION Bruce A. Boger, Director Division of Reactor Projects - III/IVN Office bf Nuclear Reactor Regulation Date of Issuance: August 5, 1992

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS ACTION (Deleted)..........................................

1-1 CHANNEL CALIBRATION (Deleted)...........................

1-1 CHANNEL CHECK (Deleted)...................................

1-1 CHANNEL FUNCTIONAL TEST (Deleted)........................

1-1 MEMBER(S) OF THE PUBLIC (Deleted).........................

1-1 OFF-SITE DOSE CALCULATION MANUAL (ODCM) (Deleted)........ 1-2 OPERABLE -

OPERABILITY (Deleted)..........................

1-2 PROCESS CONTROL PROGRAM (PCP) (Deleted)...................

1-2 REPORTABLE EVENT (Deleted)................................

1-2 YANKEE-ROWE i

I NDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCES SECTION PAGE 3/4.0 APPLICABILITY (Deleted).................................

3/4-1 3/4.1 SPENT FUEL PIT WATER LEVEL (Deleted)....................

3/4-2 3/4.2CRANE TRAVEL - SPENT FUEL PIT (Deleted).................

3/4-3 3/4.3SPENT FUEL STORAGE AREA RADIATION MONITOR (Deleted)..... 3/4-5 3/4.4 LIQUID HOLD-UP TANKS (Deleted)..........................

3/4-6 3/4.5SEALED SOURCE CONTAMINATION (Deleted)...................

3/4-7 BASES SECTION 3/4.OAPPLICABILITY (Deleted)................................

B3/4-1 3/4.1 SPENT FUEL PIT WATER LEVEL (Deleted)...................

B3/4-3 3/4.2CRANE TRAVEL - SPENT FUEL PIT (Deleted)................

B3/4-4 3/4.3SPENT FUEL STORAGE AREA RADIATION MONITOR (Deleted)....

B3/4-5 3/4.4 LIQUID HOLD-UP TANKS (Deleted).........................

B3/4-6 3/4.5 SEALED SOURCE CONTAMINATION (Deleted)..................

13/4-7 YANKEE-ROWE il

INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE LOCATION.............................................

5-1 5.2. FUEL STORAGE..............................................

5-1 Criticality (Deleted).....................................

5-1 Drainage (Deleted)........................................

5-1 Capacity (Deleted)........................................

5-1I YANKEE-ROWE iii

INDEX ADMINISTRATIVE CONTROLS SECION PAGE 6.1 RESPONSIBILITY (Deleted)..................................

6-1 6.2 ORGANIZATION (Deleted)....................................

6-1 On Site a.nd Off Site Org.nizi.on.........................

6 1 6.3 FACILITY STAFF OUALIFICATIONS (Deleted)...................

6-5 6.4 TRAINING (Deleted)........................................

6-5 6.5 REVIEW AND AUDIT (Deleted).................................

6-5 YANKEE-ROWE I.iv

I NEX ADMINISTRATIVE CONTROLS (continued)

SECTION PAGE 6.6 REPORTABLE EVENT ACTION (Deleted)........................

6-13 6.7 PROCEDURES AND PROGRAMS (Deleted).......................

6-13 6.8 REPORTING REQUIREMENTS Annual RepeFt 16 Unique Repomrtn Reiuircments.

V Spccla1 Rep9tc_.* (jD te)..............................

6-17 6.9 RECORD RETENTION (Deleted)...............................

6-18 6.10 RADIATION PROTECTION PROGRAM (Deleted)...................

6-19 6.11 HIGH RADIATION AREA (Deleted)............................

6-20 6.12 PROCESS CONTROL PROGRAM (PCP)

(Deleted)..................

6-21 6.13 OFF-SITE DOSE CALCULATION MANUAL (ODCM)

(Deleted)........

6-22 YANKEE-ROWE V

1.0 DEFINITIONS

-This section is not applicable to a facility-with all of the spent nuclear fuel stored at an Independent Spent Fuel Storaae Installation (ISFSI).

(Paue 1-2 has been deleted).

aThe i cabi termstiofo thisr setrrni~

appear i

ai""- 4 z

ryean r

4.1 A.CTION shall be tho-se addition,3l rcgulrcements Specificd a:, corollar-y statementc to each principle speclfication and shall bc part of the speci i atione.

1.2 A CHANNEL CALIBRATIONI -shall be the adjustment. a: ncccsarty, of the c-hannel output sueh th-at it respond: with the nccc::ar-y rangc and acceurac to.nw valuc: of the parameter which the channel mon~titor.

Tlhcp CHANNEL CALIBRATION shall cncompas: thc estirc ehanncl includingg the allarm.. and/or trip funstiorns, and shall 4nclude the CHANNEL FUNCTIONAL TEST., The CHA,NNEL CALIBRA.TION may be performed by any serie: of-ncquential, over-lapping or total channel Atpnq nunh thAt t~hp rntirr' P.AnAP ig calihnatodr 1

11.

%Lcr

• It 11

4.

'~

behaior urin opcatln by observation. Thie dctermination shall:

include, where possible, comparison of the channel In;dication andior status with othcr indication: andlor statu: derived from independent instrument channel: measurlng the same Parameter.

rV.ANNFI FUNIljalnlAI TrC 1.1 A C14NNEL FUNCTIONAL TEST shall hc the injec~tion of-a s~imul tcd zgnal ito th cAnnel a: lo to th prmr r.o P

practicabble tveiyOPERABILTY including alarmi and/or trip 1.6 EMRBRER(S)

OF THE PUBLIC &hall includc all persons who arc not 114 L

J L

~

YANKEE-ROWE weua

-w----

y'~ ass:ute wit te lat urgy does not include employee: of the utility, it: cotaco:,o vendors. Also excluded from thi: category are person: who enter the sitc to seryiec equipment or to make deliveric:s. Thi: category doe: include

- I I

-I - -.

1-1

1.0 DEFINITIONS (Continued) persons w:ho usc portions of the site for recreational,.occupational or o-ther purposes.v 1.6 The OFF SITE-DO0SE CALCULATION MANUAL (ODCM) contains the mcthodology and parametcrc uccd in the calculation of, off si-te doses, resulting from radioactive gaseous and liquid effluents, in the Galcul-Ation of gaseous and liquid effluent monitoring alarm/trip setpoints, and 4n the conducPt of thc Environrncntal Radiological Monitoring Program.

The QDCM also contains (1) the Radioactive Effluent Controls _and Radiological Env.ironmental Monitoring Programs r-equired by Section 61.8. and (2) descriptions of the information that should be inc-luded 4p thc Annual Radiological Environmcntai Opermatinq and AnnHUal. gadioactivc Effluent Release Reports rzequired by Speci4ficntation 6.8.2.a and 6.8.2.b.

1.7 A system, subsystem. train, component or deyice shallbe OPERABLE of have OPERABILITY when it is capable of performing its specified cafcty function(s).

!mplicit in this de-finition -shall be the, assumnption that all necessapy att~endnt 4nstpumcntation, controls.

electric jiower, coolng or seal w:ater. lubrication or-other auxiliary equipment that ape required 4fo the sYstem. subsystem.

train, component or deyice to pepformý its safety-function(s) ape also capable of performing their necessar-y support function(s).-

nnfROUSCONrA01nn fnP~AQU (PCP) 1.8 The PROCESS CONTROL PROGRAM contains the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or-simulated wet sollidd wastes, Will be anccomplished in such a way as to assure compliance with 10 CPR P-art-s 20. 61. and 71, state Fegulations, burial ground requirmets and other sequiremcnts, governing the disposal of solid radioactive, 3-9 A4 REPORTABLE EPENT shall be any of those conditions specified in 10 Amendment No.

151 1-2 YANKEE-ROWE

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3/4.1 SPENT FUEL PIT WATER ILEVEL LIMITING CONDITION FOR OPERATION 3.1.!

At least 14 feet of water-shall be maintained -ever-the tep f iffadiated fuel assemblies seated in the setoago ran1,.

3.1-.2 Mt least 5 feet of water-shall be maintained ever-the top of irrndicitzd fuel assemblies while the fuel assembties amo not seated in the retmge feclw.

3I=.3 Irradated fuel assemblies may be ftranspoted in or-out of the sepnt fuel pit, withc the mininmum 5 feet of water ever-the top of the irradiated fbucia.emblie, when dthy ame trunmpoit withiin a shipping and/or-tron.fer-onk.

APPLICABIMLTY

• hcnevff irradiate fuel i.semrblies afe in the spent fuel piL.

With thc requircments of the speeifieation sot satLfied, p!laoo irradiafted mpnnsia safe eondition and susped fuel handling epmmerans within the cpnt f1u-1 pit and crae oporation mith leads over thoe spent fuel pit. Rcstorc wat-er le vo to-withinit limit within 4 houre.

4-14. 11e weetr- !eve! in thez spent fuel pit chnhl be dctcr~niid to be at least itk minimumi

.reequi~reed_ depth last

oc nee per W.

dlays when irradiated fue4 assambliez arc in ti YANKE-ROWE 34-2Aind.

No. 149 3/4-2 YANKEE-ROWE

3/4.2 CRANE TRAVEL - SPENT FUE~L PIT UIMITING CONDITION FOR OPERATION 3.2

-Lea& in emle'S's of 900 pounds shell be prhibtcd-fm tmf'o -eye -L pn f~ i emeept fefr the7 n.peae

a.

SpentV Fuz Stofage I Buiding rof hatehes,

b. Spent fuel inspeetion atrind

e.

Ekuel handling equipment,

d.

Sp ent -fu~1rokl

e.

GfA eompeneits and assoziated lifting dovo~

fi.

ShiPPing aindr tronc-fber cazkz: 80 Gtono mnWm go Avght on Ywtd AMe Crane,

g.

Shipping cak lines;,

h:

Volume reduction equipmneft.,

h Gak-Act -^j pdand APP1L!cArlry EM ith; fuel assemblies intz pnt Rue Pit.

'.Vit tendeqiden.fts of thO GbOvz sPeifieafiens noet satisffied, plac the crane lead wna ae.niin 4.2.1 Leads in xemes of 900 pounds shall be proevoned fromn tfaveling ever-the spent Mue pit by admflin~itfafive eontrol, cxcept that the:

a.

Spent Feel Stefage Building roof hateh, spent fuel mnskpectin stand, fiefhndin

. __ eat; spent tui rnzkc, spent L asefficcblY nondstruefive iest eoauinment-Pn4U PFt 'i-ua-0 9in easr iffeff, v'olume feamouin ceIuioment. ca.scomoono n rirnr

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YANKEE-ROWE

/3 Amd. No. 149 r_-,

3/4-3

13/4.2 CRANE TRAVEL - SPENT FUEL PIT SURVEILLANCE REQUIREMENTS (Continued)

b.

Phe Yzlume radueition equipment, the sepnt futl inspeetion -stand; and the spent fuel nzcembly ncndc niuetive test euipment shall be pre-vented from Lrnvling ever-fuel assemblies in the spent fuel pit by aiministietivc eantroI. The Yoluxne rodueticn, neupmnt (when transpofted by the j'JAga Crnne), the shipping canic linerc, the e9AEse t do%% pad;, and the chipping andler-trnferý WASk shall be prevznted -from traveling over-fuel ascemblies inte spent fuel pit by adfrinistrative eontrel and by the Yarfd Area Crane trayel limnit sw.itoh,..hieh pro. ent movement of these loads beyond the seuthern end ofthew -A-Ae load path. The eask hateh c-vor-and the cock compenents and asseoiated lifting deviee shall be prevented frm traving ever spent fuel assemblies stored in "h sepnt fuel pit by adnulnistrative eontrol and by the

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components and associatod lifting devieec may be all owed over-the ripent fe assemblies in the shipping and/or tWranser cok when used in conjinction with ec I-t 2%-'

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travwel n g ever-fuel assembhice in the spent fuel pit by admaipistrative eonterl.

YANKEE-ROWE 344Ad o 4 3/4-4 Amd. No. 149

3/4.3 SPENT-FUEL STORAGE AREA RADIATIO-H MONITOR UMITING CONDITION FOR OPERATION 3.3 Radiation lovels in the spent fuel stemge ame shaHl be monitored by a fired reAdatio monitor With an audible alarm set at <5 MAW'h or-2 times background, whiehoeri APPLIGARILITY When handling ifmdiated fuel, sentfol rodo or-seurzo.

With fixod radiation monitorn c

~ipnt inepcrablc, place irradiated comonen n fe enditien and hupn nf ol hadling operations within the spent fuel-pit and craneo3peratons mith lea&c ever-flo spent fuel pit until the mfoflitor iS retwnied to GPERABL~ st ats

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X/4.4 IQUIID-HOI D-UP TANKS UNMITING CONDITION FOR OPERATION

~

3 Iaia~v ea1rn flenuulnc Hn nn-fnf--t - fn-sufrounded-by AMrdie or-Ymln eapabbk oeef heldin' tS tnk conten%~ or-that docs not have~ a tank overflow connocted to the liquid radwaste trvotment ayotem, shall be APPLIGAflILLTY 3

A--.t- -nII t im lirat% without Mzay. take aetion to suspend all-adffitiefns of ruioeative material to h WAnk Within 48 houmre rduco the tank contents to within the ling! and dezribc tb "efct. 4eadin cti odt in te xt AnnuAl R2AdioActivnF Uffilmait ThArpn-I t;1L pmureunt E0 SpeiKC1Cfleadl rmab MI., dL*

I Mi1115! _. Ittz~ell M.H.1MHN!!ý

- ~ ~

~

~

wfL' %L" WLJ nc~~t4&

p S currunded by linenc, dikes, or walls eapable of holding the tEankcnenfea shall be detcrwivdad to be 'within mlhM Abovz.' limit-by analyzing a repr-eseaaflive sample of the tnnk'c eontents at leart enco per-7 days-when radio aefive materiab wec being addo to the ta&.

.YANKEE-ROWE

./-

mnmn o 5 4-4-4 The aumaw ot radioac ve fnateFi dMontainAA in

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-3/4-6 AgEM Amendment No. 151

ssion timomefs pre lieusly r-a__.__l to eefe flu_-)--At4east-eaee mdioael.: ve atefials.2 VAth it half gfeatep t AR 3()dWJS (BXelading Hydfegen-3ý,ý

-TR Anwfe

-etheFulmn9as.

YANKEE-ROWE 314-7 eaffelgeq ofthe above H:ff d.Sý Mftg ýu Ply t-A at the-hwpeney desedbed bejew, selffees A 4r.

I I

364.5 SFAI F NTAMINATION LIMITING CONDITION FOR OPERATION

--Eaeh sealed seuf!ee eentaining ffidi0aeflYa raatefial-eýýýý Fniefoeufieer of bet-a an"r41 eFoeudes-of alpha p

ernitting-matefý hwll be fiee fn...

Ut

.--- ttm eF A ppi BMP-4!

At all times, AGNE-G-Nif Eaeh sealed

1. .-:-adon in em 9 of the abeye 1--:

be immediately m4ffid-sam"n fiem use an&

b Either-deeentmainated fu A ---

ired or

2.

Disposed of in -aeee..--.

Fdoneevi-I --.

SS.'On UNGH REQUnupýuRgS 44.1 Test RemiFementg Em-eli.ef the A w4equký,ed reufees shall u-----ed-fer ne lieensee, e

b.

G4hff pepsens-spe-eifieffilly by-Gemm-:__

state.

4le test motbod shall haye a deteetien sensit-i-Vity of at 1-es st-o.oo rMereeu

-BEEPe, A C 0)

3/4.5 SEALED SOURCE CONTAMINATION SURVEILLANCE REQUIREMENTS (Contlinued) 4-(Cefttinued

U

-~r' nOurf'e r1'WS ME' In~i "CA A-f to another-Iizenzc unless tested within the previous 6 fnonths. Seeled sourees tmnnsfcrid without a ee~dfifcvt.. indkafing the last test date -shall be te~fed prior to bzing plaeed inte use.

A D'3IU1 1CDOfl flaH D nrerarci nni Ufl1fl1ttzI to tIi (nmm1,-rn purcmn ttogie poincati t

a1S nuolP;it eezfd surce Iceiaz ostowtrA'cil thA ui~~~onw~~

m1 rrýio wa T~u mzoun7

ý_ T. r. mwhýnrtniit-w YANKEE-ROWE

/8 3/4-8

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS FOR THE DEFUELED TECHNICAL SPECIFICATIONS (Pages B3/4-1 through B3/4-7 have been deleted)

The 13ASES eentan&

in Scctien 3/4, but in speeifleefiens din the sueeeedine mees summr aeeeefflae wit IOF 5.6

-e

&ie the reasens fer thc SPcoificati re not port of these Teehaici ME,

314.0 APPLICABILITY BASES The speeifieations of this section previde the general fequirements applieable tc each of the Limiting Conditions for Operation and Sunreillane Requircments within Scetion 311.

24a1 This speeification defines the applicability of each specification in termis of apecifd

.Applieability conditions and is provided to delineaite spccifieally when cach SPecificationk isf applieable.

M.0. This specification defines these eonditions neeessar; to constitutoeemplimneewt the terms of an individual limniting Condition of Operation-and asseite.AGMflN requiemen.

M.0. This specification provides that entr; into a speeified Applicability conditien must be mawe yith (a) the full eemplement ef required systems...uipnt and copofnenets OPERpABLEy and (b) all other-parmnete r, as speeified in the Limiting Conditions for Operation being mnet without regard for-allowable deviations and out of serWiepeiin contained in the AC9TION -Statements.

The intent of this provision is to ensure that aetivities are not initiated with eiter-required equipment or-systenis inoperable or-other-speeified limnits being exeeeded.

1A0& This speeifieation establishes the requirement that survilicracc must be performfed during the speeified aplieable eonditions for whieh the requrements of-the Limiting Conditions for Operaition apply unless otherwise stated in an individual gureillane Requirement. The purpose of this specification is to ensure thW surveillances are pcrformed to verify the operatienal status of systems and-components and that paraometr are within speeified lim~its to ensure safe operation of the facility when the plant is in at specified A-pplicability condition. gurveillance Requirements do net have to be perfmormdwhen the facility is in a condition for whieh the requifrcmonts of the associated limitig Condition foi Operation do not apply unless otherwise specified.

4:Q.1 This specification establishes the limtit for-which the specified time interval fo Surveillanee Requirements may be extended. It permnits an allowable extension-of the.A normal urwyeillanee inter-Ad to fheilitae survteMillane scheduling and consideration of plnt operating eenditiens that may not be suitable for-conducting the surveillance; e.g., transien eaonditions or-other-angoing surveillance Or maintenance activities. It is net intended that this provision be used repeatedly as a convenience to extend YANKEE-ROWE 8/-

83/4-1

3/4.0 APPLICABILITY BASES (Continued) the sufveillanee intew'Las beyond that speoified. -The limitation of 9pftifieatien 4.0.2 is based on engilecming~judgement and the recognition that the moest probable iestilt of any V n part11uamr. sur_ Xmiaee bcinig perfrmied is the verifleatlon of eonformanee with the Surveillance Reguifements. Thin prevision is sufficient to ensure that the reliability ensure through sufycifll nce ectivities is not signifeantly-degraded beyond that obtained from the spceified surveillance interval.

1.03 The provsienio of thin speeifleation set ferth the eriter-ia fer-determination Ot compliance with the O1PERA.BRTY faircments of the Limiting Condifions o Gperation. Under-this eritcr-ia, equipment. systens or-eemponents arm assumced to be OPEILA3.BL if the associatcd surveilanee aetivites have been satisfaceteriy performed within the specified tie intorval. Nothing in thi provision is to be constnied as definling equipment, syntefm OF eempanents OPE3RA.BLE when suchitcms mr found or-known tcb inepemble although still mneeting the SUr-:Ilanee Requirements.

4&A Th~e speeification ensurce that the survcillanee aetivitles~ o...,,ialed with a Ufimiing Condifieon for Opemdtin have been perfomed within the epecified time interval prior-to cntry intO 6peeifled Appieability eonditiene. !Me intent of this provision is to ensure that surveillanee netivities hmv been satisfiwterily damoetrted on aurrnt bsna eurdt meet the QPERA.BE6FTY requirements of the limaitin ConditionfrOeain Under-terms of this specifleation, the appliable surveillance aetivitice nmut be performd within the stated sumvillance interval pr-ior to placing or ireturning the syctem of equipment YANKEE-ROWE B/-

B3/4-2

3/4.1 SPENT FUEL PIT WATER LEVEl BASES The restriation of 14 fiet of water-cover for-irradiated Muo assemblies seated in stompg raeks, ensures that sufflierAn water-depth is available to remove 98% of the approximately 30%

iodine gap aetivity teehased from the, rupture of an irradiated fuel assenMby. The mi~nimfum water-depth is eonsistent with the assumptions ef the aeoident analysis-.

The rcotricticn of 5 fee! of water cover-for-kradiated fuel assemblies not seated in the sterage ruckse nsurcs that the exposurm to pcr~znncl in the spent fuel pit area during fc mo-vements will be maintainted ALARA.

Mthen sperA fuel is Uunsnpefted in of: eut ef the spent fuel pht witin a shipping andlor transfer-cask; shielding for-personnel expesuro is pro-vided by the eask system. The eask YANKE-ROE 834-3Amd.

No. 149 11-

-I

_ý i.w.

YANKEE-ROWE B3/4-3

3/4.2 CRANE TRAVEL - SPENT FUEL PIT RA*F*

Tihe restrietion on movement of lea&s in exeess of the nominal weight ofa fuel assembly ever fuel assemblies in the &pent feel pit ensures that in the event this lead is "ropod (1) h aetvity feleaso will be limited to that contained in a sin&l fuel assembly, and (2) mny possible disiortion of the fuel in the storagc raks will not reult in a er-itieel arr-ay. This assumption is eonsistent wit the aeti-vity Meease assumned in the accident analysis.

Handling of the preent Spent Fuel Storage Building reef hatehes under atdmqiiStr~ativo central will assuro safe handlin of the reef hatcer. The restriction of fniovcrnnt of' th spent fuoel inspeetien stand, the spent fuel assembly nondestructive test equipment, the cask hatch covcr-, the volumeg reduetion equipment; and the shipping eask liners Gvor spent fuel ensure that thoe itemas cannot bc dropd an spent fue. Dropping any one of these itemfi from its maximum height 'wilt not rseult in loss of integrity ef the spent fuel pit floenr Ha~ndlin~g of the fuel handlin equipment for infrequent ma~intenlance under adininistrative control will ensre thc. -.

4Afc handling of any fuel handling eomponents.

The use of the Yard Crane single failure proof main hokq or-the aumiliax-y hook in accor-dance with NUREG 0612 and TS 34-.2 provides assurance that a failure rosixtin ina lead drop is net credible and will not result in the shipping andler. tranSfer cask, the cask set dewn pad; the cask hatch eo.. r-, and the cask compenents and associated lifting deviees having an adverse effect on the spent fuel pit or-the hrradated fuel in the spent fAM pit. The restriction en moevement of the shipping Rnd/OF trasfer-cask, the cask set down peAd the cask hatch eover-, and the cask components and lifting de-viecm Nfurter ensures that these itemas cannot be dropped en spent fuel in the spent fuel pit storep racks. The uso of the Yard Crane single failur proof main hook, or-the auxiliay hook in accordance with PAJMiG 0612 and TS 311.2 ensures that the cask hatch cover-and tahe eask compenent an associated lifting de-vices, which are permnited oyer-the spent fuael in the cask, cannot be dropped en the spent fuel. Tho safe lead path is established tw support the defence in depth approach to safety concerning he&vy loa&s over the Spent fuel pit. Deviadeohe from or changes to the safe load path shall be performned in aecordance with approvod written procedures which have been reviewed by an Independent Safety Re-viewer-and approved by ev.

n ss eVI~f*.pnal *Ungf, b

ungr e

Amd.

No.

155 B3/4-4 YANKEE-ROWE

3/4.3 SPENT FIIE[ STORAGE AREA-RADTATION MONITOR BASES Opefability mquir-ements for-the Spent FMe Stompo Ara Radiatio~n Monitor-sad* 10 GCR

70.24. "Gritieality Aczident Requirements," wideh mandatas the need for a R-adiatien

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flxx tA rxmwa-l M

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Co l poeified axnounts. Using radiation deteetion equipmcnt that can aefiyate an audiblo alar I ignal in the sepnt fuel stoage area enue ea notifioation of exeessively high radiation is Frowdedi te inelvidtusi ifleveiv in fue1 nandling opewation or-other-oetivitice in the Spent Fuel Stempg Building.

Me intent of Speeifiesfien 3/1.3 is to pretcct plant personnel and ensurm public healthad safcty by requhirig a Radiation Mcnitcring 85stem to be OPER.ABLE mhen rrLqdatod fe assmblies~, eeatrel reds, and seurees in the Smcnt Feel Pit wre foeved. Tnoneambla Aixe rwnauon manirnrint coninment reauir iictrnTi t rtrc tixt rartintrnn mnnituwini

eallinment toOPRH

-nn YANKEE-ROWE 8/-

-i r

B3/4-5

3/4.4 [IQUID HOLD-tUI ANMKS BASES Resricting the quantity ef "adiaetive znnteral eentained in the eutdoer-tanks proiey rncSUranc that in the cvnt of an uneentrzlkd rclease of the twnks eontents, the reswting eeneentwaion at the neamst ptnbk water-supply and the nearest surffaz wnter-suplyin unrcstfieted mra would be less then the lintits of 10 G17R 20, Appendix B, Table E6.

Column 2. The limit opplies to eaeh tank individually.

Tanks ineluded in Whs Spoeifleation am~ these outdeep tanks that eentain rndiaetiye liquid; ame not surrounded by liner-s, dikes, or-walls eapable ef holding the tank eontzntsý,

z o

hiave t&Ak ever-flo'ws and suffounding arm drains eonneeted to &-liquid radwaste treatment systeft YANKEE-ROWE 8/-

B3/4-6

314.5 SEALEO SOURCE CONTAMINATION BASES The limitation en remov:able contamunation fer 50urcm requifing leak tzsting, nldn alphax emitters, is based en 10 GFR:7039(e) limits for-plutonium. This limifin wl ensure that leakage fiom by pr-odoct; sourcoe, and speeiel nueloczr matefiail sourmc will not xemeed allowable intnk values. Scaked seurees amc elassifed into three gieups according to their use with sw-.cillanee rcquirements oemmensamto with the praobbility of damage to at souree in that group. These sourees whieh amc frogucatly handled wre requied to be toted more often than these which are not. Sealed sourees which amc eantintiously oneleftd withinf

a.

v in'U.~..A~l t*l.*

spofee V

tmnf *c.%rn..%

menit**bn oru-ur be incor deviees) arm eonsidered to, be sterod and nooad not be tested unless dthy amremove~d from h ahielded mechanism.

YANKEE-ROWEB/4-B3/4-7

5.0 DESIGN FEATURES 5.1 SITE LOCATION The Yankee Nuclear Power Station is located in the town of Rowe, Massachusetts, three-quarters of a mile south of the Vermont-Massachusetts border as shown in Figure 300-1 of the updated Final Safety Analysis Report.

The site is at the bottom of a deep valley along the Deerfield River on thve southcast bank of ghcrman Resepyoip. Thc gcographical coordinates of-the Gcnterline of the reactor eontainment structure- (Vanpor Contir)rea I

_.4&I..

A

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-A 5.2 FUEL STORAGE "r'L -

rL*.

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u'vv am2 a

V maintained with a center to center distance between fuel assemblies plac-ed in the storage racks to ensure -a-k-,

igyni4.

+n.~

n ng-with *unborat*d water. The k. of 0.95 inludes A

-AnAPPr"TAZ nrnt.n"+

+4 A,.n.,.v~,

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The spent f-ue!

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4nnes.

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n~n4~ni4

+.44I 4 An rs

.,41 A mrimm ofpa5it' spemit fuel assemerec from th0 f

uel mblle4 A maximum of 540 spent fuel assemblies from thp YAnIkpp Nfu-lpr Power Station are stored in dry casks within an Tndrnpndpnt Spent Fuel Storaae Installation (ISFST).

Amd No.

150 YANKEE-ROWE W 720 F;G' 42" a.0- 1 :k

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6.0 ADMINISTRATIVE CONTROLS This section is not applicable to a facility with all of the spent nuclear fuel stored at an Independent Spent fuel Storage Installation CISFSI.).

(Pages 6-42 through 6-22 have been deleted) fkW%&n4:srrati-.-

eentrelo ar: the %ori.tten rules, ordere, i.-.truetjr, pr~zduz.,

relieizei, praetiees, and the designatien of autherities and rxe)p~en-R ~ibiii

---

.~ the manzinxgirat to obtain assuran: eE oafety and quality of rmzirnteanee eE a nuelear feeilit,-. These eenr.Lrle shall be ad~hered to.

Gz~s;6 The Deeennisaienin; Haftagr shall be reposj:ible for avera!!

feeilit,- operatier. anel ehlta delegate in writing~ the suecessizr. to thisJ respensibi11tb-durin;-his beianr.:z 6.122 Ir 11nttere relabing to the operation of the plant and to thA T:-.Hieet Speefieabti:,

the Peeenwxxieining Mana.ger shallrzp9t to and be direably xespenzible to the Viee President aE Y&Wee-A1er:

B.zxc e:trie Czempey.

6.z2 ORCAMIVZ.T-14 6.241 An en. site and an eff site organi~zabie ohmll be-esbablished n

-r iir rn-i-i

-raniizt shall z:ij neluzl thez przitierks ' F

£..n:s et &Utfterit,-

zznibiity, an811 nz-ez ohal be srnilez and dettnce1 Erem thez flliost non& z;mcnt

!a:1 trieugh intermezdiate levels to and ineluding all operating~ organization pz.sitionsz~

Tha..

relationships ahall-be daeuvented and upd;detZd, a a;;repriate, in thez fenr el ergar~iatiornal ehzxts.

These organization*! eha.rts shall be deeumenrted-in the.-,

~

~

rrteer eveL+/-a?:Zra rear

+/-iytr ln reulear eafzty.

4This individual:

ahall take any measures needed to ensure aeptable per.f.~.an...: of the staff in eprattirg, so tha:t eentireded rmuelear safety is assaured.

ex T~here ehall be an ind-ividual management penitlen in the on eite organizatier. having overall rspensilbilit,- Eer YANKE-ROW 6-1Amendment No. 154 sake operation. of thez plant, hz,'nhe zsb:l have-eentre~l eve-r these-on elte reseurecn neeessary for 1anf:

operation and naintenene-

__f 0.- viant.

I -

4 6-1 YANKEE-ROWE

6.0 ADMINISTRATIVE CONTROLS (Continued)

e.

Although the indrviduals "a%`o trrt-the opera t taEf and theso who zarry cut the quality toouranzo furnctiowe may roport to the appropriabeto xcnaer. they ahall have stufficiont organizational froocion to bo irAopznlornt Ereo oporatir.; pressuros.

Although radiain prot l may roprt to the a"Lrepriac m.anager on site, for matters roalting to radiological health and saafty of employees and the puhijo. the aito/rtoir rnfanaer shall haveo diroot a.....

to that on sito ini"vidual bavying ovraill responsibility for oaf e operation of the plant.

Radiation orotootia;. nroraor-el ohm!!l hav: the authorityf to ocaso any work aotivity wheni oafcty in joopardizcd or in the workor radialogloal event of urmrezozacay GA.;

The0 racntl-erpanitation Ofla.LJ noe sunocot to etro to*iovan~.

a.

Beeh on duty ehift shall bo compricod of at leaot tho minimum ahift; cre: eeop3itien ohowm in Vale 6.2 -1.

b.

At loeast one individuial uaalifiod to staemd watch in the control roo0M (an Equipment ~Oprator or Cartifiod Fuel Hlandler) shall be* in tho conttrol ream when fuel. is ix the spent fuel pit-.

e. A qualified Radiation rrotaction individutal shall ho orn eitt durin~g fuol hiandEirc oporationas.

di.

Arn ircdividual rrnlified in radiatior. proteotion.

procadurco shall ho orn site when fuel is in tho opent fuel pit:

Operating crew pcrooticl tand nrdito pretoction proooderoz fill thia reqjuiremont.

e.

All fuel komdlir.; oporationsobafl ha dirootly oukporvisod by-a Ceztifiod Fuel Hlandier who h-an mne-oener

• 71.

Amendment No. 153 6-2 YANKEE-ROWE

6.0 ADMINISTRATIVE CONTROLS (Continued) vau d neez zzzr. aeziztza YANKEE-ROWE I

6-3

6.0 ADMINISTRATIVE CONTROLS (Continued)

HiflEMi!

GH;FV GREW GGHPOOCITIHt heft414

1. t Shif~t erew eampeoltier rray be leao than the ~dim r uirefflentzi Fra pr ae t ti~me not to ezte:e 9 heurig in ozrder to azzzaacz U

tz.taz1 aJzzaa: at ziut-; att ar

rnbzr

aa vianc tom 7ea ar~zaUl J1t r

ez aspitien t ihr i

~i~az fTbaC21 YANKEE-ROWE64 I

6-4

6.0 ADMINISTRATIVE CONTROLS (Continued) 6.3 PAGILITY CTP.IT OUlTrIGzq.TG?;

G.3.1 gaeh mzrer cE tile facility mnagemeit auperviz-eery

~~1 meet-or eiczze3 the:

Biiuiu1f~in E 7.NGI 18.1 1i)7i Eer zemparable posi~tions, esceept fzr the Radiatitn Preteetian Ma~nager w.ho zihall a-le meezt the minimum qualfieatjens -"

Regulater-; Guide 1.8, PeFl-izz 1.

6.

retraiin-,~3rprin tann r;au ~

h

~ii~

~n UCnpr~-z-taiing PV

_;rrn

~ti~

uni ztaf zal ~z e -ze eh bzrizft 0-nd!

rzrzn~tir.~of Ceetien 5.5 :fmszwg

?Y1R.11'Y1 6 -

DTMLfTED e

tU___

-idj r.~e beenr zdz1eted YANKEE-ROWE65 6-5

6.0 ADMINISTRATIVE CONTROLS (Continued)

6. 6 IFPORSrAJLE EV3RT-TACTIDII Gs6.6 The following azetier..

shall be tziha Ear HBPRqrIMLB BVBFPG

a.

The Gen~aion.9hra1 be matifiecl and a report submrnitte purauant toz the requireme:nts ef 10 CPR 50.73, sn~

b.

Bee~h RlEPORTABLE BMFPU shall be reviewed by an Ia.ndpendzrnt Safety~ Rev-iewer and1 the zzzsulte of thle rev'iew sihall bez subazittee5 to the ;nde1pendent Review. anel Atidib Czrittee

- (IlIAC) and the DBzcr=missiining Hmmerr.

YANKEE-ROWE 61 Pages 6 14 threttob 6 15 have been delete 6-13

6.0 ADMINISTRATIVE CONTROLS (Continued)

• U 1U.atIUrllK iaxjuinnznirrc 10 Clii 50.4. ';he reporting requiremen~ts of 6.6A,1.8. a nd 6.8.3 speeifizzdtzns."

6.zu. 8.1 P~vrt Am.u&! reparts eavering t~h cet~i4ties ef the

__i -t e-icbz1 belew for the preyiems yen zluil YAP siubm.itted prior to llareh 1 zf eaeh year.

Repo:rts required on aft am~ual basis~ she!! inelqlz;



.

ZflZI2 22 22.2 

-t I1

reeeiv.ing exptree greater than 100 mremý',-

a~nd thezir ted ma rm empcau~re neeeding t: werle and jobb Anserviee &nspeet~zr., reetine mainbertanee, opeeial YANKEE-ROWE 61 q4te fellewing identified reports shaii be submaibbed persuemt-be 6-16

6.0 ADMINISTRATIVE CONTROLS (Continued)

The des: assigment to vazrieue dluty fu ztiens~ mty be estim~ates bmee onpacket dai:L:ter, q%%, er Filma ha~;ZA fnesreerits.

total dose: need nab he aeau~nt:ed f

-r-. the agregate, at least 80% o.f the whel: bed~- dese reeei;e frem emteinvil eazurces ehall mcazn1tcz.ii ajor worle-Etketins.

b.

ArW, other unit unique reports, r:-eirzeI en an annua!

6..2 Unqu Reoting Requrementsi~zuz uiar interprtations. -nrd analyaiaz ef trend &E the tresult ef the Thadielagieal Btinvrmnmbnal flerdtering Program Eer the repertieg peried. The m.terial pr:-.-idiz shall be zeenziabet with the ebleetiveg eutlinzz1 in (1) the GPGi, and (21 Czetiens I;.B.2,~ P;,B,3, sadz r;.G. OE Appendimc to 10 cmR 50.r epertien of the unit 1,.zing-the precAeus shall be submitted befere Hay 1 of eeh ryear. The report -ehall inelude a eiumr of the qeantitie: aE -Ladeati--e fremn the unit. The material previdzei aball be (1) znzsistent with the objeet...e ettblir.e in the GPQ! an:

PGc, and (2) in eenfernanez with -10 GM 50.36a and Czetion r:.B.i zf AppzndL I te -10 eFR 50.

6.8.3 fpeeia l Reprte -Speeial vrert. shall1 be sftrdtted pursuarnt te 10 Gr1 59.4 it the time peried speeifiŽ-A Eer eselt report. Thes: repars shall-be submittedi eeyeriny the the app1ilable referenee aneeifieatieri.

YANKE

- OWE

-17Amendment No. 151 6-17 YANKEE-ROWE

6.0 ADMINISTRATIVE CONTROLS (Continued)

a.

6ealed Gettre DEaEtTEe in zf 1-imit8, Speeificablen 3.5.

6. 9 DELEPEEM Amendment No. 155 YANKEE-ROWE 6-18

6.0 ADMINISTRATIVE CONTROLS (Continued) 6.1g D IIDIPTICN PROTECTIONG)

PflOORAU LA.+/-U.2 t'reeeures for personnel radiation ivreteetkon ahalal h prepairre cons~istent %fith re iremzrts af -10 GE'T Part 20 and eperations yrefelving persn~nel radIiatilen expesurczi YANKEE -ROWE 61 6-19

6.0 ADMINISTRATIVE CONTROLS (Continued) 6.161 1110G11 RADIATIOU MLDA 6 9:1. 1 Paagr:rPh 20.903, "Gautian Signs, lnabel, Signals, n Gentralew.'

in lieu af the lentral devieel er "atlm eignaVy required boy Pt'ragraph 20.203(z) (2), eaebh Ugk radlatiert area in whieb the intensity ef radiationr. 1:1000 arftremhr or less ohall be bh rieaded and eemzpiewuzusly posted ae a hih radiationr aL-ea, and entranee thereto oball be eenretrlled b requirfr;ing ta~za a Radintien WrLe Pranit (nRM)

Ar. indiv-idua ar;rp in~dividuals peAratted to eniter sueh

rWPRn nhall be pvn;vided with one: er more of-the fellaeirng.

a.A radiation ranit:Urkn deie w...hieh aarntir.iously indleates the razdiatien dese rate in the arat.

b.

A radiation mni....t......

deice w....hieb eentirnueuzly integrates the radiationn dan Late in the area, an alarma whewn a preset integrated-dose ie reeeiyed. Bntry in~to uahil areas wi~t thia rmnitcring devieea r~ay bec tm~de after t-e done Late leve:l in the "ara has been etablished and lperjrsennl have been zm-.e kole~,.Ldgeablel e

themft e.A fladiatien Pro teeti on --ea1efi d indi4.i uial 1i.:.,

qa1lified in radkatien preteetian lpreeed-res), w:ith et radiatien dose rate menitarixi; da-:iee, whoin rsarnsible fer prev.iding positive eentral a.-:: the aebiv.ities w.ithin.

the area and who will l;:rfarr radiatiar. sur-;aillanne at the Erequeney speeified in the iiMr.

The surveaillance Erequeney will be establiehed by-the Rladiatiorn Protection Thea bove: preeedure eke!! aloe apply to eaeh high radiation a.rea in wi;Lih the intensity of-radiationn isz granter than 1:000 rare.-.h*. In ad~dition, icaked-deers shall be prc~avde ta preveant unau.therized entry into uatth ara:: and 'he a aball be m.ain~tanerd u~nder the. a1zirnietrative eent-rel.fE t S hift:

Superv:isor an.

duty and,'ar the fRadiation Prrteetlen UH.nager.

Rladiation I'rnteatiar. paaareanl shall be exemapt frate the lWr issuark:

pretaetien duties, providing t~hey are Eellewing plant radiatiar preteetian preeedures far entry into hieth radiation raa YANKEE-ROWE 62 6

6.0 ADMINISTRATIVE CONTROLS (Continued) 6 A2 va1 Ghmanez to the i'cr.

shall be daezumeat~

shall be retained ean%~ reerea rev&

fs~ nerter-eeifsied t-YD pr eppnd"

1)

Suffieient in~e~vatien-+e sz upport the change; tzgether

..ibh bhe aprpit emalyse e r evaluatiz justif~-yr.; the ehnge(al, and

21)

A deberzwn~tier. that the ehanxge will xaintain the overall eanfermmee of the aelidAiedc. waote praiuzet to exizitin; -

uiz.nbe OF Eedeet, abate# er other

b.

Shall bem ffeetive Tndependnt Sa~fety Revi uz~iioin~ui uznaoecr a

eee ntr i-I--

(or a deignee) rea+o-h YANKE-OWE -21Amendment No. 155 B (R~eeer4 sp a.

aeuentmenis This Aaeumentava-van an-all sant-n ____ -1 6-21 YANKEE-ROWE

6.0 ADMINISTRATIVE CONTROLS (Continued) 6.13 OF SITEr DOSE~ eA~eUh.,-Q HMatT (0,*mttlyv VI~.

b-~fteftfes to-fte 9vi

0.

bn*

-ne aeettmzntzz ana reeerzas et rev~zws erkreien viairyeApzda~

(flzzr~flztntia).

thia doeumnntatizr. ahai! e ntain.

1)

SfEfie~nt-4nf:....atien toa uppert the eh-ar.az teaether withi the apprepriata ara).1-ez or evaluatean

2)

. 1ztazz.iration that tha I!1

fl1 r1rt-4

,- i-'--

levelz of the ra icaetive effluent eantral required~ by 10 CIFA 2941Q6, 99 eMi 9:9, 15036a and A~ppendix 1 to 10 Gcm 50 and1 not advearsely itmpaet the aeeurey-r eeirhility eE effittent, dose, or Be4t"aint eaal1Aul=A'irAf%

C. z~naiarazem zrraecetve krtr andi ae::;tanzz b~y -u Xn~~a~~a..~~.zz~~a "ea:f:i~.

r taa;z:1o h

Daz3.ar11r.

Uana:r Car a Izainaa).

C.

Shall ea aubmitted to the Cari--2aar

.n

-I-- he fem Of et b ad :

OE 0part OL en aurra-tb 16th the Tx~wa1 pmdiefttiv lffu flealsa fl-r for the perie3 af Utha... :rt i;- w:hieh any-elicui 11-the: OD! w:as made.

rbaeh ehange abet!! be identifiedY hb-,

qaxZdl;ng in the margin ef the affeeted p~age, aecar1l

ýi eaing~ the. area of the page bhat %p:as ehanged,1 and ahanll ir4L-t: lt~hea te (e.g., meaath,'~ar h

f.

iip;ienented.

eag a

YANKE-ROWE6-2

~Amendment No. 155 guat--Eymv the ehanaels). an I ---

ý __Iýl I

6-22 YANKEE-ROWE

ATTACHMENT II Revised License and Technical Specification Pages License 3

5 Appendix A i through v I-1 3/4-1 Bases Cover Page 5-1 6-1

Amendment No. 145 9/4/1992 and to the rules, regulations and orders of the Commission now or hereafter In effect;, and is subject to the additional conditions specified or incorporated below:.

(1)

Maximum Power Level The licensee Is not authorized to operate the reactor. Fuel may not be placed in the reactor vessel.

(2)

Technical Specifications The Technical Specifications contained In Appendix A, as revised through Amendment No. 155, are hereby Incorporated in the Ucense. The licensee shall possess and maintain the facility In accordance with the Technical Specifications.

(3)

Physical Protection Amendment No. 156 3/13/2002 The licensee shall fully Implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revision to 10 CFR 73.55 (51 FR 27817 and 17822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plan, which contain Safeguards Information protected under 10 CFR 73.21, is entitled, 'Yankee Nuclear Power Station Security Plan,"

which includes the "Contingency Plan" and the "Guard Training and Qualification Plan,* with revisions submitted through June 28, 2001.

Changes made in accordance with 10 CFR 73.55 shall be Implemented in accordance with the schedule set forth therein.

(4)

Fire Protection Amendment No. 144 8/20/1992 The licensee shall implement and maintain In effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for the facility and as approved by NRC Safety Evaluation Reports dated March 15, 1979, and as supplemented October 1, 1980, and August 27, 1986, subject to the following provisions:

The licensee may make changes to the approved Fire Protection Program without prior NRC approval if these changes do not reduce the effectiveness of fire protection for facilities, systems, and equipment which could result in a radiological hazard, taking Into account the decommissioning plant conditions and activities.

D.

This license is effective as of the date of issuance and authorizes ownership and possession of this facility until the Commission notifies the licensee in writing that the license is terminated. The licensee shall:

1. Take actions necessary to decommission and decontaminate this facility and continue to maintain this facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition; and

2. Conduct activities in accordance with all other restrictions applicable to this facility In accordance with NRC regulations and the specific provisions of this 10CFR50 facility license.

FOR THE NUCLEAR REGULATORY COMMISSION Bruce A. Boger, Director Division of Reactor Projects - III/IVN Office of Nuclear Reactor Regulation Date of Issuance: August 5, 1992

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS ACTION (Deleted)..........................................

1-1 I CHANNEL CALIBRATION (Deleted).............................

1-1 CHANNEL CHECK (Deleted).................................

1-1 1 CHANNEL FUNCTIONAL TEST (Deleted).........................

1-1 MEMBER(S) OF THE PUBLIC (Deleted).........................

1-1 OFF-SITE DOSE CALCULATION MANUAL (ODCM)

(Deleted).........

1-2 OPERABLE -

OPERABILITY (Deleted)..........................

1-2 PROCESS CONTROL PROGRAM (PCP) (Deleted)...................

1-2 REPORTABLE EVENT (Deleted)................................

1-2 I YANKEE-ROWE i

I NDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCES SECTION PAG 3/4.OAPPLICABILITY (Deleted).................................

3/4-1 3/4.1 SPENT FUEL PIT WATER LEVEL (Deleted)....................

3/4-2 3/4.2CRANE TRAVEL -

SPENT FUEL PIT (Deleted).................

3/4-3 3/4.3 SPENT FUEL STORAGE AREA RADIATION MONITOR (Deleted)....

3/4-5 3/4.4 LIQUID HOLD-UP TANKS (Deleted)..........................

3/4-6 3/4.5SEALED SOURCE CONTAMINATION (Deleted)...................

3/4-7 BASES SECTION PAGE 3/4.OAPPLICABILITY (Deleted)................................

13/4-1 3/4.1 SPENT FUEL PIT WATER LEVEL (Deleted)...................

B3/4-3 3/4.2CRANE TRAVEL - SPENT FUEL PIT (Deleted)................

B3/4-4 3/4.3 SPENT FUEL STORAGE AREA RADIATION MONITOR (Deleted)..... B3/4-5 3/4.4 LIQUID HOLD-UP TANKS (Deleted).........................

B3/4-6 3/4.5 SEALED SOURCE CONTAMINATION (Deleted)..................

B3/4-7 YANKEE-ROWE I

Ji

INDEX DESIGN FEATURES SECTION MAGE 5.1 SITE LOCATION.............................................

5-1 5.2. FUEL STORAGE....................

5-1 Criticality (Deleted).....................................

5-1 Drainage (Deleted)........................................

5-1 Capacity (Deleted)........................................

5-1I YANKEE-ROWE iii

TNDEX ADMINISTRATIVE CONTROLS SECTION PAE 6.1 RESPONSIBILITY (Deleted)..................................

6-1 6.2 ORGANIZATION (Deleted)....................................

6-1 6.3 FACILITY STAFF OUALIFICATIONS (Deleted)...................

6-5 6.4 TRAINING (Deleted)........................................

6-5 6.5 REVIEW AND AUDIT (Deleted)................................

6-5 YANKEE-ROWE 1v

INDEX ADMINISTRATIVE CONTROLS (continued)

SECTI ON PAGE 6.6 REPORTABLE EVENT ACTION (Deleted)........................

6-13 6.7 PROCEDURES AND PROGRAMS (Deleted)........................

6-13 6,8 REPORTING REQUIREMENTS (Deleted).........................

6-17 6.9 RECORD RETENTION (Deleted)...............................

6-18 6.10 RADIATION PROTECTION PROGRAM- (Deleted)...................

6-19 6.11 HIGH RADIATION AREA (Deleted)............................

6-20 6.12 PROCESS CONTROL PROGRAM (PCP)

(Deleted)..................

6-21 6.13 OFF-SITE DOSE CALCULATION MANUAL (ODCM)

(Deleted)........

6-22 YANKEE-ROWE V

1.0 DEFINITIONS (Continued)

This section is not applicable to a facility with all of the spent nuclear fuel stored at an Independent Spent Fuel Storage Installation (ISFSI).

(Page 1-2 has been deleted).

Amendment No.

YANKEE-ROWE 1-1

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS This section is not applicable to a facility with all of the spent nuclear fuel stored at an Independent Spent Fuel Storage Installation (IsFsI).

(Pages 3/4-2 through 3/4-8 have been deleted)

YANKEE-ROWE 3/4-1

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS FOR THE DEFUELED TECHNICAL SPECIFICATIONS (Pages B3/4-1 through B3/4-7 have been deleted)

5.0 DESIGN FEATURES 5.1 SITE LOCATION The Yankee Nuclear Power Station is located in the town of Rowe, Massachusetts, three-quarters of a mile south of the Vermont-Massachusetts border as shown in Figure 300-1 of the updated Final Safety Analysis Report.

5.2 FUEL STORAGE A maximum of 540 spent fuel assemblies from the Power Station are stored in dry casks within an Spent Fuel Storage Installation (ISFSI).

Yankee Nuclear Independent YANKEE-ROWE I I 5-1 Amd No.

6.0 ADMINISTRATIVE CONTROLS This section is not applicable to a facility with all of the spent nuclear fuel stored at an Independent Spent fuel Storage Installation (ISFSI).

(Pages 6-2 through 6-22 have been deleted)

YANKEE-ROWE 6-1 Amendment No.

ATTACHMENT III The following 7 pages of text will be included in the YDQAP

LIQUID HOLD-UP TANKS LIMITING CONDITION FOR OPERATION The quantity of radioactive material contained in any-outside tank that is not surrounded by liners, dikes or walls capable of holding tank contents, or that does not have a tank overflow connected to the liquid radwaste treatment system, shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.

APPLICABILITY:

At all times.

ACTION:

With the quantity of radioactive material in any outside tank exceeding the above limit, without delay, take action to suspend all additions of radioactive material to the tank.

within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit and describe the events leading to this condition in the next Annual Radioactive Effluent Release Report.

SURVEILLANCE REQUIREMENTS The quantity of radioactive material contained in any outside tank that is not surrounded by liners, dikes, or walls capable of

• holding the tank contents shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.

YANKEE-ROWE

SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of

>0.005 microcuries of removable contamination.

APPLICABILITY:

At all times.

ACTION:

Each sealed source with removable contamination in excess of the above limits shall be immediately withdrawn from use and:

1. Either decontaminated and repaired, or

2.

Disposed of in accordance with commission Regulations.

SURVEILLANCE REQUIREMENTS Test Reauirements -

Each of the above required sealed sources shall be tested for leakage and/or contamination by:

a.

The licensee, or

b.

other persons specifically authorized by Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

Test Freguencies -

Each category of the above required sealed sources shall be tested at the frequency described below.

a.

sources in use (excluding startup sources and fission detectors previously subjected to core flux) - At least once per 6 months for all sealed sources containing radioactive materials:

1.

with a half-life greater than 30 days (excluding Hydrogen 3), and

2.

in any form other than gas.

YANKEE-ROWE

SEALED SOURCE CONTAMINATION SURVEILLANCE REQUIREMENTS (Continued)

(continued)

b.

Stored sources not in use - Each sealed source shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months.

sealed sources transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

Reports -

A Special Report shall be prepared and submitted to the commission pursuant to 10 CFR 50.4 annually if sealed source leakage tests reveal the presence of greater than 0.005 microcuries of removable contamination.

YANKEE-ROWE

FACILITY STAFF QUALIFICATIONS Each member of the facility management/supervisory staff shall meet or exceed the minimum qualifications of ANSI 18.1-1971 for comparable positions, except for the Radiation Protection Manager who shall also meet the minimum qualifications of Regulatory Guide 1.8, Revision 1.

REPORTABLE EVENT ACTION The following actions shall be taken for REPORTABLE EVENTS:

a.

The commission shall be notified and a report submitted pursuant to the requirements of 10 CFR 50.73, and

b.

Each REPORTABLE EVENT shall be reviewed by an Independent safety Reviewer and the results of this review shall be submitted to the Independent Review and Audit Committee (IRAC) and the Site Manager.

REPORTING REQUIREMENTS The following identified reports shall be submitted pursuant to 10 CFR 50.4.

The'reporting requirements of the following three sections are in accordance with Revision 4 of Regulatory Guide 1.16, "Reporting of operating Information - Appendix A Technical specifications.

Annual Report Annual-reports covering the activities of the unit as described below for the previous year shall be submitted prior to March 1 of each year.

Reports required on an annual basis shall include:

a.

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and job functions, (a) e.g., operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), and waste processing.

The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the whole body dose received from external sources shall be assigned to specific major work functions.

b.

Any other unit-unique reports required on an annual basis.

2 Unique Reporting Requirements

a.

Environmental Radiological Monitoring The Annual Radiological Environmental operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year.

The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Prog ram for the reporting period.

The material provided shall be consistent with the objectives outlined in (1) the ODCM, YANKEE-ROWE

and (2)

Sectibns IV.B.2, IV,B,3, and IV.C of Appendix I to 10 CFR 50.

b.

Annual Radioactive Effluent Release Report The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year.

The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be (1) consistent with the objectives outlined in the OCM and PcP, and (2) in conformance with 10 CFR 50.36a and section IV.B.1 of Appendix I to 10 CFR 50.

3 special Reports Special reports shall be submitted pursuant to 10 CFR 50.4 within the time period specified for each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a.

sealed source leakage in excess of 10 CFR 70.39(c) limits.

RADIATION PROTECTION PROGRAM I

Procedures for personnel radiation protection shall be prepared consistent with requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposures.

YANKEE-ROWE

HIGH RADIATION AREA I

Paragraph 20.203, "Caution signs, Labels, Signals, and Cont'rols."

In lieu of the "control device" or "alarm signal" required by Paragraph 20.203(c)(2), each high radiation area in which the intensity of radiation is 1000 mrem/hr or less shall be barricaded and conspicuously posted as a high radiation area, and entrance thereto shall be controlled by requiring issuance of a Radiation work Permit (RWP).

An individual or group of individuals permitted to enter such areas shall be provided with one or more of the following:

a.

A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area, and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate level-in the area has been established and personnel have been made knowledgeable of them.

c.

A Radiation Protection qualified individual (i.e.,

qualified in radiation protection procedures), with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and who will perform radiation surveillance at the frequency specified in the RWP.

The surveillance frequency will be-established by the Radiation Protection Manager.

The above procedure.shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem/hr.

In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the key shall be maintained under the administrative control of the shift supervisor on duty and/or the Radiation Protection Manager.

Radiation Protection personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, providing they are following plant radiation protection procedures for entry into high radiation areas.

YANKEE-ROWE

PROCESS CONTROL PROGRAM (PCP)

I changes to the PCP:

a.

shall be documented and records of reviews performed shall be retained as specified by YDQAP Appendix D (Record Retention).

This documentation shall contain:

1) sufficient information to support the change together with the appropriate analyses or evaluation justifying the change(s), and

2)

A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of federal, state, or other applicable regulations.

b.

shall become effective after review and acceptance by an Independent Safety Reviewer and the approval of the site Manager (or a designee).

OFF-SITE DOSE CALCULATION MANUAL (ODCM)

I Changes to the ODCM:

a.

shall be documented and records of reviews performed shall be retained as specified by YDQAP Appendix D (Record Retention).

This documentation shall contain:

1) sufficient information to support the change together with the appropriate analyses or evaluation justifying the change(s), and

2)

A determination that the change will maintain the level of the radioactive effluent control required by 10 CFR 20.106, 40 CFR 190, 10 CFR 50.36a, and Appendix I to 10 CFR 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b.

shall become effective after review and acceptance by an Independent safety Reviewer and the approval of the Site Manager (or a designee).

c.

shall be submitted to the commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

YANKEE-ROWE PROCESS CONTROL PROGRAM (PCP)